1. Field of the Invention
This invention relates in general to the field of pressurized water nuclear reactors and in particular to the field of coolant water piping arrangements between the reactor vessel and the steam generators of a pressurized water nuclear reactor.
2. Description of the Prior Art
In very simplified and general terms, pressurized water nuclear reactors include primary and secondary flow systems. The primary system is usually a closed system using light water as the reactor coolant and the nuclear moderating fluid. The light water reactor coolant is heated upon flowing through a nuclear core which is contained within a pressure vessel. The heated reactor coolant is then ducted through a steam generator where it transfers heat to another fluid within the secondary system. The fluid in the primary system is, of course, arranged in heat transfer relationship, within the steam generator, to the fluid in the secondary system. Finally, the reactor coolant flows to a main reactor coolant pump which pumps the coolant back through the pressure vessel repeating the above-described flow cycle.
The fluid within the secondary system most often is also light water which is converted into steam within the steam generator. The steam is used to power a steam turbine which is mechanically connected to an electrical generator which in turn converts the power originally derived from the nuclear core into electricity.
A typical commercial nuclear power plant may contain four separate primary flow systems each of which is associated with its own separate secondary flow system. Each primary flow system is generally termed a "primary coolant flow loop", or more simply, a "flow loop". The portion of the flow loop from the reactor coolant pump to the pressure vessel is usually designated as the "cold leg"; while the portion from the reactor vessel to the steam generator is usually designated as the "hot leg".
In a flow loop of the prior art, the cold leg includes piping connecting the main coolant pump to the reactor vessel. The hot leg includes piping connecting the reactor vessel to the steam generator. A "crossover" pipe leg comprises the piping connecting the steam generator to the main coolant pump. Each of the described piping legs are series flow connected but are otherwise physically removed or separated from each other. Thus, in the highly unlikely event of a break or fracture in any of the legs, some of the primary core coolant would be vented from the reactor vessel and would be released to but contained within the reactor containment building.
A failure of the exposed cold leg primary coolant piping between the pump and the reactor vessel, such as a large break (greater than six (6) inch internal diameter pipe) loss-of-coolant accident could result in extensive damage. Such an accident could cause c ore uncovery by the reactor coolant resulting in high peak clad temperatures of the fuel elements in the vicinity of 2000.degree. F. to 2200.degree. F. This in turn could cause extensive damge to the nuclear core and thereby release radioactive fission products to and within the reactor containment building. It is to be noted that the described accident and the described results are mere possibilities-not probabilities.
There are other accidents which can also be envisioned notwithstanding the low probability of the occurrence of the same. For example, the crossover leg in the nuclear reactors of the prior art, which transports the coolant flow from the steam generator to the main coolant pump suction or inlet could also fracture or break. Since the crossover leg piping typically drops to a level several feet below the top of the nuclear core before turning vertically upward to the reactor coolant pump suction thereby forming a loop seal (because of its physical geometry) which inhibits hot leg venting of steam in the unlikely event of a small loss-of-coolant accident (a break of a pipe having an internal diameter less than six (6) inches). However, such an accident could theoretically uncover the core to the extend that is is blanketed by steam but not liquid water, resulting in peak fuel clad temperatures in the range of 1100.degree. F. to 1300.degree. F. Temperatures of this level can cause fuel damage which may make the core unstable which could result in the release of fission products to the reactor coolant supply system or to the containment building.
Accidents of the type described could cost millions of dollars in recovery expense plus two or more years of lost generation of power costs. Obviously, a relatively expensive and undesirable result.
Still another undesirable aspect of the flow loops of the prior art is again concerned with the loop seal crossover leg. In the event of a loss-of-coolant accident the loop seal acts as a trap for the condensate from the steam generator. The level of the condensate will gradually build up until a slug of water overflows from the trap and hits the main coolant pump impeller. Thereafter, steam will impact upon the impeller until the level of condensate again builds up, overflows, and causes the impeller to be struck by another slug of water. Such alternating action continues and can cause excessive vibration resulting in the necessity to shut down the pump. Loss of the pump following a loss-of-coolant accident is undesirable because the pump could be used to circulate steam water mixtures thereby providing an additional and diverse means of cooling the core and preventing core damage.
Accordingly, a primary object of the present invention is to provide a pipe loop arrangement which eliminates the possibility of a large loss-of-coolant accident and the attendant consequences thereof as a result of a failure of the cold leg piping.
Another object of the present invention is to provide a pipe loop arrangement which prevents the core from being uncovered by liquid water during a small loss-of-coolant accident resulting from a failure of the cold leg piping.
Still another object of the present invention is to provide a coolant pump and pipe loop arrangement wich upon a loss-of-coolant accident will continuously supply a mixture of steam and water to the impeller of the main coolant pump for continued circulation of the steam-water mixture during the period of time following the accident.
A further object of the present invention is to provide a primary system pipe loop arrangement having the cold leg piping concentrically mounted within the hot leg piping.
A still further object of the present invention is to provide a primary system pipe loop arrangement which minimizes the length of piping to reduce the piping pressure drop.
An even further object of the present invention is to provide a primary system pipe loop arrangement which allows for increasing the efficiency of the reactor coolant pump.
An even further object of the present invention is to provide a primary system pipe loop arrangement which eliminates the need to have a loop seal crossover leg.